Michael F. Simpson

Electrolytic Reduction of Spent Nuclear Oxide Fuel as Part of an Integral Process to Separate and Recover Actinides from Fission Products

Abstract Bench‐scale tests were performed to study an electrolytic reduction process that converts metal oxides in spent nuclear fuel to metal. Crushed spent oxide fuel was loaded into a permeable stainless steel basket and submerged in a molten salt electrolyte of LiCl–1 wt% Li2O at 650°C. An electrical current was passed through the fuel basket and a submerged platinum wire, effecting the reduction of metal oxides in the fuel and the formation of oxygen gas on the platinum wire surface. Salt and fuel samples were analyzed, and the extent of fission product separation and metal oxide reduction was determined.

A Description of the Ceramic Waste Form Production Process from the Demonstration Phase of the Electrometallurgical Treatment of EBR-II Spent Fuel

Abstract The electrometallurgical treatment (EMT) process has been designed and developed for stabilizing sodium-bonded, metallic fuel into two high-level waste forms. This process has recently been successfully demonstrated with irradiated EBR-II fuel at Argonne National Laboratory-West. Part of the EMT process is to immobilize fission-product-bearing waste salt, which results from electrorefining, in a ceramic waste form—a glass-bonded sodalite. The sodalite is formed by hot isostatically pressing salt-loaded zeolite at temperatures up to 850°C and pressures up to 100 MPa. The specific unit operations that comprise ceramic waste production include steps for salt grinding, zeolite drying, blending salt and zeolite and glass frit in a v-blender, and consolidating the powders in a hot isostatic press. The results of testing these unit operations with irradiated salt from the EMT demonstration are summarized and include some preliminary characterization of the final irradiated ceramic waste form created by this process.

Actinide recovery experiments with bench-scale liquid cadmium cathode in real fission product-laden molten salt

Abstract This article summarizes the observations and analytical results from a series of bench-scale liquid cadmium cathode experiments that recovered transuranic elements together with uranium from a molten electrolyte laden with real fission products. Variable parameters such as the ratio of Pu3+/U3+ in the electrolyte, liquid cadmium cathode voltage, and feed materials were tested in the liquid cadmium cathode experiments. Actinide recovery efficiency and Pu/U ratio in the liquid cadmium cathode product under variable conditions are reported in this paper. Separation factors for actinides and rare earth elements in the molten LiCl-KCl/cadmium system are also presented.

Development of Computational Models for the Mark-IV Electrorefiner—Effect of Uranium, Plutonium, and Zirconium Dissolution at the Fuel Basket-Salt Interface

Abstract The electrochemical processing of spent metallic nuclear fuel has been demonstrated by and is currently in operation at the Idaho National Laboratory (INL). At the heart of this process is the Mark-IV electrorefiner (ER). This process involves the anodic dissolution of spent nuclear fuel into a molten salt electrolyte along with a simultaneous deposition of pure uranium on a solid cathode. This allows the fission products to be separated from the fuel and processed into an engineered waste form. A one-dimensional model of the Mark-IV ER has begun to be developed. The computations thus far have modeled the dissolution of the spent nuclear fuel at the anode taking into account uranium (U3+), plutonium (Pu3+), and zirconium (Zr4+). Uranium and plutonium are the two most important elements in the system, whereas zirconium is the most active of the noble metals. The model shows that plutonium is quickly exhausted from the anode, followed by dissolution of primarily uranium, along with small amounts of zirconium. The total anode potential as calculated by the model has been compared to experimental data sets provided by INL. The anode potential has been shown to match the experimental values quite well with root-mean-square (rms) values of 2.27 and 3.83% for two different data sets, where rms values closer to zero denote better fit.

Modeling the Pyrochemical Reduction of Spent UO2 Fuel in a Pilot-Scale Reactor

Abstract A kinetic model has been derived for the reduction of oxide spent nuclear fuel in a radial flow reactor. In this reaction, lithium dissolved in molten LiCl reacts with UO2 and fission product oxides to form a porous, metallic product. As the reaction proceeds, the depth of the porous layer around the exterior of each fuel particle increases. The observed rate of reaction has been found to be dependent only upon the rate of diffusion of lithium across this layer, consistent with a classic shrinking core kinetic model. This shrinking core model has been extended to predict the behavior of a hypothetical, pilot-scale reactor for oxide reduction. The design of the pilot-scale reactor includes forced flow through baskets that contain the fuel particles. The results of the modeling indicate that this is an essential feature in order to minimize the time needed to achieve full conversion of the fuel.

Development of electrorefiner waste salt disposal process for the EBR-II spent fuel treatment project

The results of process development for the blending of waste salt from the electrorefining of spent fuel with zeolite-A are presented. This blending is a key step in the ceramic waste process being used for treatment of EBR-II spent fuel and is accomplished using a high-temperature v-blender. A labscale system was used with non-radioactive surrogate salts to determine optimal particle size distributions and time at temperature. An engineering-scale system was then installed in the Hot Fuel Examination Facility hot cell and used to demonstrate blending of actual electrorefiner salt with zeolite. In those tests, it was shown that the results are still favorable with actinide-loaded salt and that batch size of this v-blender could be increased to a level consistent with efficient production operations for EBR-II spent fuel treatment. One technical challenge that remains for this technology is to mitigate the problem of material retention in the v-blender due to formation of caked patches of salt/zeolite on the inner v-blender walls.

Nuclear Fuel Reprocessing

This is an a submission for the Encyclopedia of Sustainable Technology on the subject of Reprocessing Spent Nuclear Fuel. No formal abstract was required for the article. The full article will be attached.

A Computational Model of the Mark-IV Electrorefiner: Phase I―Fuel Basket/Salt Interface

Spent driver fuel from the Experimental Breeder Reactor-II (EBR-II) is currently being treated in the Mk-IV electrorefiner (ER) in the Fuel Conditioning Facility (FCF) at Idaho National Laboratory. The modeling approach to be presented here has been developed to help understand the effect of different parameters on the dynamics of this system. The first phase of this new modeling approach focuses on the fuel basket/salt interface involving the transport of various species found in the driver fuels (e.g. uranium and zirconium). This approach minimizes the guessed parameters to only one, the exchange current density (i0). U3+ and Zr4+ were the only species used for the current study. The result reveals that most of the total cell current is used for the oxidation of uranium, with little being used by zirconium. The dimensionless approach shows that the total potential is a strong function of i0 and a weak function of wt% of uranium in the salt system for initiation processes.

Diffusion Model for Electrolytic Reduction of Uranium Oxides in a Molten LiCl-Li2O Salt

Abstract In this study, a diffusion-based kinetic model essential for design and operational analysis of spent nuclear fuel reduction has been developed. The model considers the cathode side of the system to be rate limiting and deals with diffusion of lithium metal through the basket loaded with uranium oxide (UO2 or U3O8). Faraday’s law was implemented into the model to observe the electrochemical effect on the model. Solutions with different conditions are developed, and detailed results are presented. These solutions were compared against experimental bench scale data. At high operating current conditions (I > 0.8 A), the model fits the data well. The fitting resulted in estimated effective lithium diffusion coefficients for high and low void fraction UO2 crushed fuels of 8.5 × 10−4 cm2/s and 2.2 × 10−4 cm2/s, respectively. The effective diffusion coefficient for U3O8 is estimated to be 8.6 × 10−4 cm2/s. In some experiments, a porous magnesium oxide basket was used for containing the U3O8. It was estimated that the lithium diffusion coefficient through this magnesia basket is 3.3 × 10−5 cm2/s.

COMPUTATIONAL MODEL OF THE MARK-IV ELECTROREFINER: TWO-DIMENSIONAL POTENTIAL AND CURRENT DISTRIBUTIONS

Abstract A computational model of the Mark-IV electrorefiner is currently being developed as a joint project between Idaho National Laboratory, Korea Atomic Energy Research Institute, Seoul National University, and the University of Idaho. As part of this model, the two-dimensional potential and current distributions within the molten salt electrolyte are calculated for U3+, Zr4+, and Pu3+ along with the total distributions, using the partial differential equation solver of the commercial Matlab software. The electrical conductivity of the electrolyte solution is shown to depend primarily on the composition of the electrolyte and to average 205 mho/m with a standard deviation of 2.5 × 10–5% throughout the electrorefining process. These distributions show that the highest potential gradients (thus, the highest current) exist directly between the two anodes and cathode. The total, uranium, and plutonium potential gradients are shown to increase throughout the process, with a slight decrease in that of zirconium. The distributions also show small potential gradients and very little current flow in the region far from the operating electrodes.